An active neutron method for measuring the inherent neutron emission of spent fuel assemsly

Tạp chí KHOA HỌC ĐHSP TPHCM Tran Quoc Dung<br /> _____________________________________________________________________________________________________________<br /> <br /> <br /> <br /> <br /> AN ACTIVE NEUTRON METHOD FOR MEASURING<br /> THE INHERENT NEUTRON EMISSION OF SPENT FUEL ASSEMSLY<br /> <br /> TRAN QUOC DUNG*<br /> <br /> ABSTRACT<br /> An active neutron method for measuring the inherent neutron emission of spent fuel<br /> assembly is proposed. The count rate of the inherent neutron emission can be determined<br /> by changing intensity of neutron irradiating source. The practical meaning of the method<br /> is presented. Some attractive features of the method are shown.<br /> Keywords: neutron interrogation, non-destructive techniques, spent fuels, neutron<br /> measurements.<br /> TÓM TẮT<br /> Phương pháp mới sử dụng nguồn neutron để đo sự phát xạ neutron<br /> vốn có trong các bó nhiên liệu đã cháy<br /> Một phương pháp neutron chủ động để đo lượng neutron vốn có trong nhiên liệu đã<br /> cháy được đề xuất. Tốc độ đếm của sự phát xạ neutron có thể được xác định bằng cách<br /> thay đổi cường độ của nguồn chiếu neutron. Ý nghĩa thực tiễn của phương pháp được trình<br /> bày. Một số tính năng hấp dẫn của phương pháp này được chỉ ra.<br /> Từ khóa: tương tác nơ-tron, kĩ thuật không phá hủy, nhiên liệu đã cháy, các phép đo<br /> nơ-tron.<br /> <br /> 1. Introduction<br /> In nuclear material safeguards the determination of the characteristics of spent<br /> fuel assembly such as burn-up, total fissile content, amount of plutonium and original<br /> enrichment is important. These parameters are useful for establishing critical safety in<br /> spent fuel ponds and in reprocessing plants. There are some different non-destructive<br /> methods developed for fuel identification such as: acombination of active neutron<br /> interrogation and passive neutron measurement (Shulze G. and Wurz H.,1979), the<br /> spectroscope of fission product gamma radiation and passive neutron counting<br /> (Vidovszky I. et.al.,1986; Bernard P. et al.,1986), a simple passive neutron and gross<br /> gamma measurement (Phillips J.R et al.,1981), a combination of neutron and gamma<br /> measurement (Fox G.H. et al., 1987).<br /> Because neutron measurements have advantageous features such as high<br /> transparency of the assembly, easy detectability, high neutron emission of the spent<br /> fuel and favorable signal- to- background ratio. The measurement systems based on<br /> the first method have been developed and tested in actual installations (Wurz H et al.,<br /> 1990; Simon G.G, Sokcic-Costic M.,2002).<br /> <br /> *<br /> Ph.D., Centre for Nuclear Techniques<br /> <br /> 153<br /> Tư liệu tham khảo Số 43 năm 2013<br /> _____________________________________________________________________________________________________________<br /> <br /> <br /> <br /> <br /> According to the first method, the inherent neutron emission Cne is determined by<br /> passive neutron measurement and the thermal flux multiplication Mth by active neutron<br /> interrogation measurement after Cne is known. From these quantities the primary<br /> neutron emission correlating with the burn-up, the total fissile content, original<br /> enrichment of the spent fuel is obtained<br /> This paper presents an active neutron method, with changing intensity of neutron<br /> irradiating source, for measuring the inherent neutron emission Cne of spent fuel<br /> assembly.<br /> 2. The method<br /> The principle of the method is shown in Fig 1.<br /> In a given spent fuel assembly there are the inherent neutrons (Cne) emitting from<br /> spontaneous fissions and (,n) reactions. When the fuel assembly is irradiated by the<br /> neutrons of the external source the fission reactions are induced in the fissile isotopes<br /> as 235U, 239Pu, 241P. These are detected by measuring the thermalized prompt fission<br /> neutrons. Suppose that the fuel assembly is irradiated by the neutron source leaving the<br /> intensity I1, the total count rate Ct1 of detector is given as<br /> <br /> cne<br /> Neutron source Ct1<br /> with intensity I1 Cd1<br /> <br /> <br /> Ci1 Neutron<br /> detector<br /> <br /> <br /> <br /> cne<br /> Neutron source Ct2<br /> with intensity I2 Cd2<br /> <br /> <br /> Ci2 Neutron<br /> detector<br /> <br /> Fig 1. Principle of the method<br /> Ct1  Ci1  Cd1  C ne (1)<br /> Where:<br /> Ci1 - the contribution of the fission neutrons to the total count.<br /> <br /> <br /> 154<br /> Tạp chí KHOA HỌC ĐHSP TPHCM Tran Quoc Dung<br /> _____________________________________________________________________________________________________________<br /> <br /> <br /> <br /> <br /> Cd1 - the contribution of the direct source neutrons i.e., source neutron<br /> penetrating the fuel assembly without being captured<br /> Cne - the contribution of the inherent neutron emission of the spent fuel. For the<br /> given fuel assembly Cne is constant.<br /> 2<br /> Similarly, the expression of the total count rate Ct of the same detector when the<br /> fuel assembly is irradiated by neutron source having intensity I2 is given as:<br /> Ct2  Ci2  Cd2  Cne (2)<br /> 2 2 1 1<br /> The quantities Ci and Cd are similarly defined as Ci and Cd , respectively. By<br /> subtracting Cne from the total count rate, the neutron flux increase due to induced<br /> fission is obtained. The thermal neutron flux multiplication is given as:<br /> Ct1  Cne Ci1<br /> M th   1  (3)<br /> Cd1 Cd1<br /> Or<br /> Ct2  Cne Ci2<br /> M th   1  (4)<br /> Cd2 Cd2<br /> From the expressions (3) and (4) we have:<br /> Ci1 Ci2<br />  (5)<br /> Cd1 Cd2<br /> With supposing the intensity I2 is stronger than I1 and the quantity Cd2 is n times<br /> 1 2 1 2 1<br /> bigger than Cd , i.e., Cd  nCd , the expression (5) leads that Ci  nCi and<br /> <br /> Ci2  Cd2  n(Ci1  C d1 ) (6)<br /> Combining Eqs. (1), (2) and (6) result in<br /> Ct1  (Ci1  Cd1 )  Cne<br /> Ct2  n (Ci1  Cd1 )  Cne<br /> By solving this equation system, the expression for the inherent neutron emission<br /> Cne is given as<br /> nCt1  Ct2<br /> Cne  (7)<br /> n 1<br /> 2<br /> The physical nature of this method is shown in Fig.2. From eq.7 the quantity Ct ,<br /> the total count rate of the detector with intensity I2, is obtained as<br /> <br /> 155<br /> Tư liệu tham khảo Số 43 năm 2013<br /> _____________________________________________________________________________________________________________<br /> <br /> <br /> <br /> <br /> Ct2  nCt1  ( n  1)Cne (8)<br /> 2<br /> If n = 0 i.e., the neutron source is removed, so Ct  C ne . This is the very<br /> passive neutron measurement presented in [1].<br /> If n = 1, i.e., the intensity of the irradiating source is not changed. so Ct2  Ct1 .<br /> This is obvious.<br /> <br /> Ct2<br /> <br /> <br /> <br /> <br /> 2Ct1-Cne<br /> <br /> Ct1<br /> Cne<br /> <br /> <br /> <br /> 1 2 3 4 n<br /> <br /> Fig 2. The Ct2 versus the change of the intensity of the neutron source<br /> 2<br /> choosing n>1, the linear functional dependence between Ct and n is given as in Fig 2,<br /> and Cne is the very intersection point of the line and the coordinate axis.<br /> The count rate of Cd1 and Cd2 due to the direct source neutrons are determined in<br /> the laboratory [1], so n is obtained easily.<br /> 3. Conclusion<br /> The inherent neutron emission Cne and the flux multiplication Mth are two<br /> necessary quantities for spent fuel identification. The method combining active and<br /> passive neutron measurement has allowed the obtainment of these quantities.<br /> This paper presents the method determining Cne and Mth by only active neutron<br /> measurements with changing intensity of interrogating source.<br /> This method has attractive features as follows:<br /> - Calibrations for the passive neutron measurements are not necessary. Calibrations<br /> for the active measurements are simple.<br /> - The measuring instruments are not complicated or expensive.<br /> - Intensity of the interrogating source can be easily changed by readjusting the<br /> window of source.<br /> <br /> <br /> 156<br /> Tạp chí KHOA HỌC ĐHSP TPHCM Tran Quoc Dung<br /> _____________________________________________________________________________________________________________<br /> <br /> <br /> <br /> <br /> REFERENCES<br /> 1. 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